Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Kobe, Mitsuru*; Tsunoda, Hirokazu*; Mishima, Kaichiro*; Kawasaki, Akira*; Iwamura, Takamichi
JAERI-Tech 2003-016, 68 Pages, 2003/03
JAERI-Tech-2003-016.pdf:4.37MB
The 200 kWe uranium nitride fueled lithium cooled fast reactor "RAPID-L" combined with thermoelectric power conversion system that can be operated unmanned without refueling for up to ten years has been demonstrated. The RAPID refueling concept enables quick and simplified refueling, and achieves plant design lifetime over 20 years. A significant advantage of the RAPID-L design, which does not require the use of control rods - is the introduction of the innovative reactivity control systems: lithium expansion module (LEM), lithium injection module (LIM) and lithium release module (LRM). LEM is the most promisiong candidate for improving inherent reactivity feedback. LEMs could realize burnup compensation. LIMs assure sufficient negative reactivity feedback in unprotected transients. LRMs enable an automated reactor startup by detecting the hot standby temperature of the primary coolant. All these systems use 
Li as liquid poison and are actuated by highly reliable physical properties (volume expansion of Li for LEM, and freeze seal melting for LIM and LRM).
Kobe, Mitsuru*; Tsunoda, Hirokazu*; Mishima, Kaichiro*; Kawasaki, Akira*; Iwamura, Takamichi
Genshiryoku eye, 48(1), p.23 - 28, 2002/01
no abstracts in English
Tsunoda, Hirokazu*; Sato, Osamu*; Okajima, Shigeaki; Yamane, Tsuyoshi; Iijima, Susumu; Kobe, Mitsuru*
Proceedings of International Congress on Advanced Nuclear Power Plants (ICAPP) (CD-ROM), 6 Pages, 2002/00
no abstracts in English
Ooka, Norikazu; Iyoku, Tatsuo; ; Takikawa, Noboru*; Shiozawa, Shusaku; *; *; *; *
JAERI-M 93-003, 26 Pages, 1993/01
no abstracts in English
Haga, Kazuo; Kambe, M.; Kataoka, H.; Otani, N.; Otsubo, A.
Acta Astronautica, 26(5), p.349 - 357, 1992/05
Times Cited Count：2 Percentile：40.75(Engineering, Aerospace)None
*
PNC TN9600 90-002, 106 Pages, 1990/04
The author attended the 6th Symposium on Space Nuclear Power Systems held in Albuquerque, New Mexico, USA on January 1989. The purpose of attending the symposium is to motivate further development of fast reactor technology within the framework of frontier research activities of PNC. (1)More than 600 participants, including nuclear scientists and engineers from the Soviet union, France, England and Japan and a host of US government space and energy officials, attended. Presentation including : (a)SP-100 program to develop a hundreds of kilowatts of space nuclear reactor and its apprication to lunar and martian surface. (b)Multimegawatt program to develop megawatts of space nuclear reactor and nuclear propulsion. NASA, DOD and DOE are collaborating on above programs. (2)From the Soviet delegarion, programs on thermionic reactor for earth orbit and nuclear propulsion for manned Mars mission vehicle are presented. (3)French delegation detailed the recent R&D program of 20
200 kWe of space reactor ERATO.
Otani, Nobuo*; *; *; Haga, Kazuo*
PNC TN9410 89-145, 98 Pages, 1989/10
Core physics of small reactor was examined as a part of conceptual design study of space reactor which is an application of transportable reactor. The design requirements were a fast spectrum reactor using nitride fuel and lithium coolant. Firstly, characteristics of typical uranium core and plutonium core was compared by means of one-dimensional calculation using simple sphere model. Followings were revealed from the comparison of both calculation results. (1)Reactivity loss of uranium core in ten years is smaller than that of plutonium core. (2)Shorter lifetime of plutonium core is due to 
decay of Pu-241. Hence, plutonium core must be designed to compensate higher fuel degradation. (3)Installation of thermal neutron absorber between core and reflector region is effective to extend lifetime of plutonium core. Secondly, study to optimize the design of enriched uranium core was performed as a parameter of core configuration, fuel composition and core size from the point of reactivity adjustment. Attention was focused on criticality of fresh core, reactivity degradation, sub-criticality and reactor shut down margin. The results showed that some safety margin can be obtained although it was less than that objected. A 3-dimensional Monte Carlo code was partly used in the analysis. It was revealed that a more superior code in simulating core configuration is necessary.
*; *; *; *
PNC TN9410 89-081, 136 Pages, 1989/04
We studied the concept of the small and medium-sized FBR as one of FBR development strategy. This year, we investigated the characteristics of reactor performances of the small and medium-sized FBR and the concept of innovative plant. Main results are as follows; (1)The characteristics of reactor performances of the small and medium-sized FBR (100 to 600MWe) was clear. (2)One problem of reactor performances of the small and medium-sized FBR is the reduction of the burn-up reactivity loss, because the smaller the reactor size is, the larger the burn-up reactivity loss is. (3)The reactivity coefficients of the small and medium-sized FBR, which is smaller size than 30kWe, is considerably different from large FBR. (4)According to quasi static approach, to have inherent shutdown capabilities, the linear heat rate should be reduced (below 250W/cm at 100MWe core) and the reactivity worth of one control rod should be limit to about 0.4%
k/kk'. (5)We established the concept of 100MWe innovative plant, making use of the characteristics of small reactor. The amount of core structure per power size of this plant is equal to Large FBR (1000MWe). Further, we should examine the amount of whole plant including BOP. Because of lower core pressure drop and linear heat rate, the capability of removing decay heat of this plant is superior to Large FBR. Because the fuel cycle cost of this plant is inferior to Large FBR (1000MWe), we should design high burn-up core.
*; *
PNC TN9420 87-001, 107 Pages, 1987/06
A research on the current perspective of overseas fast breeder reactor development programmes has been made in order to discuss the orientation of Japanese first demonstration FBR as well as the optimum procedure for the introduction of the following FBRs into the existing economic network in Japan. The most part of the report was devoted to the development programme of western Europe. In addition the efforts of the united States were briefly reviewed.
*; *; *; *; *; *; *
PNC TN941 84-169, 172 Pages, 1984/12
A design study was conducted in order to establish the concept of the FBR plant eliminating the secondary sodium loop. Conditions for realization and effectivenesses for cost reduction were also studied for this plant. The main topic for understanding this plant concept was recognized as to clearify "the influence of sodium-water reaction to the reactor core", broken out in the steam generator. Discussions were mainly focussed on the reactor core, the steam generator, the containment vessel, sodium-water reaction product relief system and so on, which were supposed to be especially important for this plant concept. Following items were recognized. (1) Total image and concept of the FBR plant eliminating the secondary sodium loop. (2) Influence of the sodium-water reaction product, especially hydrogen gas, to reactor core and limit of the water-leak rate for core damage. (3) Countermeasures for reduction and elimination of these influences. (4) Concepts of the safety map for sodium-water reaction of this plant and requirements for water leak detection systems. (5) Necessity of the duplex tube type steam generator from the view point of property protection. (6) Reduction effect for the amount of materials and construction cost by adopting this concept of plant. A overall process for design study was experienced throuqh the activities of this work.
*; *; *
PNC TN951 84-04, , 1984/07
PNC-TN951-84-04.pdf:3.61MBPNC-TN951-84-04REV1.pdf:2.98MB
no abstracts in English
